1. Field of the Invention
The present invention generally relates to a zirconium based alloy usable for the formation of strips and tubing for use in nuclear fuel reactor assemblies. Specifically, the invention relates to new technology that improves the in-reactor corrosion and/or the in-reactor creep of Zr—Nb based alloys by an essential and critical final heat treatment. The invention was applied to Zr—Nb based alloys that were developed by alloying element additions and exhibit improved corrosion resistance in water based reactors under elevated temperatures.
2. Description of the Prior Art
In the development of nuclear reactors, such as pressurized water reactors and boiling water reactors, fuel designs impose significantly increased demands on all of the fuel components, such as cladding, grids, guide tubes, and the like. Such components are conventionally fabricated from zirconium-based alloys commercially titled ZIRLO, corrosion resistant alloys that contain about 0.5-2.0 wt. % Nb; 0.9-1.5 wt. % Sn; and 0.09-0.11 wt. % of a third alloying element selected from Mo, V, Fe, Cr, Cu, Ni, or W, with the rest Zr, as taught in U.S. Pat. No. 4,649,023 (Sabol et al.). That patent also taught compositions containing up to about 0.25 wt. % of the third alloying element, but preferably about 0.1 wt %. Sabol et al., in “Development of a Cladding Alloy for High Burnup” Zirconium in the Nuclear Industry: Eighth International Symposium, L. F. Van Swan and C. M. Eucken, Eds., American Society for Testing and Materials, ASTM STP 1023, Philadelphia, 1989. pp. 227-244, reported improved properties of corrosion resistance for ZIRLO (0.99 wt. % Nb, 0.96 wt. % Sn, 0.10 wt. % Fe, remainder primarily zirconium) relative to Zircaloy-4.
There have been increased demands on such nuclear core components, in the form of longer required residence times and higher coolant temperatures, both of which cause increase alloy corrosion. These increased demands have prompted the development of alloys that have improved corrosion and hydriding resistance, as well as adequate fabricability and mechanical properties. Further publications in this area include U.S. Pat. No. 5,940,464; 6,514,360 (Mardon et al. and Jeong et al.) and Reexamination Certificate U.S. Pat. No. 5,940,464 C1 (both Mardon et al.), and the paper “Advanced Cladding Material for PWR Application: AXIOM™”, Pan et al., Proceedings of 2010 LWR Fuel Performance/Top Fuel/WRFPM, Orlando, Fla. 09/26-29/2010 (“technical paper”).
Mardon et al. taught zirconium alloy tubes for forming the whole or outer portion of a nuclear fuel cladding or assembly guide tube having a low tin composition: 0.8-1.8 wt. % Nb; 0.2-0.6 wt. % Sn, 0.02-0.4 wt. % Fe, with a carbon content of 30-180 ppm, a silicon content of 10-120 ppm and an oxygen content of 600-1800 ppm, with the rest Zr. Jeong et al. taught a niobium containing zirconium alloy for high burn-up nuclear fuel application containing Nb, Sn, Fe, Cr, Zr with possible addition of Cu. The Pan et al. “technical paper” lists Alloys listed as X1, X4, X5, X5A, but deliberately only very generally describes the actual composition weight percentages, being very vague in this regard. Pan et al. reports tensile strength, elongation and creep test data, and shows micrographs and in-reactor corrosion and oxide thickness data.
Aqueous corrosion in zirconium alloys is a complex, multi-step process. Corrosion of the alloys in reactors is further complicated by the presence of an intense radiation field which may affect each step in the corrosion process. In the early stages of oxidation, a thin compact black oxide film develops that is protective and retards further oxidation. This dense layer of zirconia exhibits a tetragonal crystal structure which is normally stable at high pressure and temperature. As the oxidation proceeds, the compressive stresses in the oxide layer cannot be counterbalanced by the tensile stresses in the metallic substrate and the oxide undergoes a transition. Once this transition has occurred, only a portion of the oxide layer remains protective. The dense oxide layer is then renewed below the transformed oxide. A new dense oxide layer grows underneath the porous oxide. Corrosion in zirconium alloys is characterized by this repetitive process of growth and transition. Eventually, the process results in a relatively thick outer layer of non-protective, porous oxide. There have been a wide variety of studies on corrosion processes in zirconium alloys. These studies range from field measurements of oxide thickness on irradiated fuel rod cladding to detailed micro-characterization of oxides formed on zirconium alloys under well-controlled laboratory conditions. However, the in-reactor corrosion of zirconium alloys is an extremely complicated, multi-parameter process. No single theory has yet been able to completely define it.
Corrosion is accelerated in the presence of lithium hydroxide. As pressurized water reactor (PWR) coolant contains lithium, acceleration of corrosion due to concentration of lithium in the oxide layer must be avoided. Several disclosures in U.S. Pat. Nos. 5,112,573 and 5,230,758 (both Foster et al.) taught an improved ZIRLO composition that was more economically produced and provided a more easily controlled composition while maintaining corrosion resistance similar to previous ZIRLO compositions. It contained 0.5-2.0 wt. % Nb; 0.7-1.5 wt. % Sn; 0.07-0.14 wt. % Fe and 0.03-0.14 wt. % of at least one of Ni and Cr, with the rest Zr. This alloy had a 520° C. high temperature steam weight gain at 15 days of no more than 633 mg/dm2. U.S. Pat. No. 4,938,920 to Garzarolli teaches a composition having 0-1 wt. % Nb; 0-0.8 wt. % Sn, and at least two metals selected from iron, chromium and vanadium. However, Garzarolli does not disclose an alloy that had both niobium and tin, only one or the other.
Sabol et al. in “In-Reactor Corrosion Performance of ZIRLO and Zircaloy-4,” Zirconium in the Nuclear Industry: Tenth International Symposium, A. M. Garde and E. R. Bradley Eds., American Society for Testing and Materials, ASTM STP 1245, Philadelphia 1994, pp. 724-744, demonstrated that, in addition to improved corrosion performance, ZIRLO material also has greater dimensional stability (specifically, irradiation creep and irradiation growth) than Zircaloy-4. More recently, U.S. Pat. No. 5,560,790 (Nikulina et al.) taught zirconium-based materials having high tin contents where the microstructure contained Zr—Fe—Nb particles. The alloy composition contained: 0.5-1.5 wt. % Nb; 0.9-1.5 wt. % Sn; 0.3-0.6 wt. % Fe, with minor amounts of Cr, C, 0 and Si, with the rest Zr.
While these modified zirconium based compositions are claimed to provide improved corrosion resistance as well as improved fabrication properties, economics have driven the operation of nuclear power plants to higher coolant temperatures, higher burn-ups, higher concentrations of lithium in the coolant, longer cycles, and longer in-core residence times that have resulted in increased corrosion duty for the cladding. Continuation of this trend as burn-ups approach and exceed 70,000 MWd/MTU will require further improvement in the corrosion properties of zirconium based alloys. The alloys of this invention provide such corrosion resistance.
Another potential way to increase corrosion resistance is through the method of forming of the alloy itself. To form alloy elements into a tubing or strip, ingots are conventionally vacuum melted and beta quenched, and thereafter formed into an alloy through a gauntlet of reductions, intermediate anneals, and final anneals, wherein the intermediate anneal temperature is typically above 1105° F. for at least one of the intermediate anneals. In U.S. Pat. No. 4,649,023 to Sabol et al., the ingots are extruded into a tube after the beta quench, beta annealed, and thereafter alternatively cold worked in a pilger mill and intermediately annealed at least three times. While a broad range of intermediate anneal temperatures are disclosed, the first intermediate anneal temperature is preferably 1112° F., followed by a later intermediate anneal temperature of 1076° F. The beta annealing step preferably uses temperatures of about 1750° F. Foster et al., in U.S. Pat. No. 5,230,758, determined the formability and steam corrosion for intermediate anneal temperatures of 1100° F., 1250° F., and 1350° F. An increase in intermediate anneal temperature is associated with an increase in both formability and corrosion resistance. U.S. Pat. No. 5,887,045 to Mardon et al. discloses an alloy forming method having at least two intermediate annealing steps carried out between 1184° to 1400° F.
Note that the prior art for corrosion improvement summarized above involves alloying element additions and different intermediate anneal temperatures, but, notably, not the final anneal heat treatment temperature. Rudling et al., in, “Corrosion Performance of Zircaloy-2 and Zircaloy-4 PWR Fuel Cladding,” Zirconium in the Nuclear Industry: Eight International Symposium, ASTM STP 1023, L. F. Van Swam and C. M. Eucken, eds. American Society for Testing and Materials, Philadelphia, 1989, pp. 213-226, reported that Zr-4 fuel rod cladding fabricated from the same ingot with final heat treatments of stress-relieved (SRA) and fully recrystallized (RXA) exhibited similar oxide thickness corrosion (see Table 1).
TABLE 1Post irradiation oxide thickness of Zr-4cladding after 1-cycle of irradiation.Final Heat4 Rod Average of the MaximumTreatmentOxide Thickness (μm)SRA12 +/− 1RXA10 +/− 1
Foster et al., in U.S. Pat. No. 5,125,985, presents a straightforward method of controlling the creep by use of the final area reduction and intermediate anneal temperature. A decrease in final area reduction decreases creep, and an increase in intermediate anneal temperature decreases creep. In different applications, the in-reactor creep can be more important than in-reactor corrosion. One such example is fuel rods containing fuel pellets coated with ZrB2. ZrB2 is a neutron absorber. When neutrons are absorbed, He gas is released which increases the rod internal pressure. In this case, creep resistant cladding is necessary so that the fuel/cladding gap remains closed. A closed fuel/cladding gap ensures that the fuel temperatures do not increase due to the formation of a He gas gap between the fuel and cladding. The new technology presented below in the Summary of the Invention will show that either the cladding corrosion or the cladding in-reactor creep may be improved by the final heat treatment.
A further issue in nuclear reactors is corrosion of welds utilized in a nuclear fuel assembly. In a typical fuel rod, nuclear fuel pellets are placed within the cladding, which is enclosed by end caps on either end of the cladding, such that the end caps are welded to the cladding. The weld connecting the end caps to the cladding, however, generally exhibits corrosion to an even greater extent than the cladding itself, usually by a factor of two over non-welded metal. Rapid corrosion of the weld creates an even greater safety risk than the corrosion of non-welded material, and its protection has previously been ignored. In addition, grids have many welds and the structural integrity depends on adequate weld corrosion resistance.
Thus, there continually remains a vital need, even in this late stage of nuclear power development, for novel zirconium cladding alloys that exhibit improved corrosion resistance and improved in-reactor irradiation creep resistance over known alloys in the field, and improved welds for holding end caps on claddings and for joining grid straps that likewise exhibit increased corrosion resistance. And, as can be seen, these cladding art patents and papers provide an extremely compact art area, where only very minor changes have shown, after extended testing, major and dramatic improvements. Thus, minor improvements can easily establish patentability in this specific area.
Accordingly, an object of the present invention is to provide Zr—Nb alloys with improved corrosion resistance and/or improved in-reactor irradiation creep resistance through the selection of a specific type combination of final heat treatment.
New technology presented below in the Summary of the Invention, and elsewhere in the specification following, will show that the in-reactor corrosion is, in part, unexpectedly dependent on the specific type of final heat treatment.